Pressurized water reactors (PWRs) for nuclear power generation facilities utilize both pumped and natural circulation of the primary coolant to both cool the reactor core and heat the secondary coolant to produce steam which may be working fluid for a Rankine power generation cycle. The existing natural circulation PWRs suffer from the drawback that the heat exchange equipment is integrated with and located within the reactor pressure vessel. Such an arrangement not only makes the heat exchange equipment difficult to repair and/or service, but also subjects the equipment to corrosive conditions and results in increased complexity and a potential increase in the number of penetrations into the reactor pressure vessel. In addition, locating the heat exchange equipment within the reactor pressure vessel creates problems with respect to radiation levels encountered for crews to repair the heat exchange equipment in proximity to the radioactively hot components of the reactor vessel. The general view has also been that the heat exchangers should be located in the reactor vessel to achieve natural circulation in those systems which may utilize this type of flow circulation.
The reduction of vulnerabilities within nuclear power generation facilities is always an ongoing issue. For example, large pipes are seen as creating the potential for a “large break” Loss of Coolant Accident (LOCA) event, and thus it is desirable to remove large pipes where possible.